The nuclear fuel material of a boiling water reactor is usually enclosed in a corrosion- resistant, non-reactive, heat-conductive container or cladding. The fuel cladding serves two primary purposes: first, to prevent contact and chemical reactions between the nuclear fuel and either the coolant or moderator if present, or both; and second, to prevent the highly radioactive fission products, some of which are gases, from being released from the fuel into the coolant or moderator or both. Common cladding materials are stainless steel, aluminum and its alloy, zirconium and its alloys, niobium (columbium), certain magnesium alloys, and others. The failure of the cladding, due to the build-up of gas pressure or high temperatures in the fuel, can contaminate the coolant or moderator and the associated systems with intensely radioactive long-lived products to a degree which interferes with plant operation.
Problems have been encountered in the manufacture and in the operation of nuclear fuel elements which employ certain metals and alloys as the cladding material due to the reactivity of these materials under certain circumstances. Under normal circumstances, zirconium and its alloys, i.e., Zircaloy-2 and Zircaloy-4, are excellent materials for use as nuclear fuel cladding since they have low neutron absorption cross sections and at temperatures below about 750.degree. F. are strong, ductile, extremely stable and non-reactive in the presence of demineralized water or steam, which are commonly used as reactor coolants and moderators. Zircaloy is an alloy of zirconium with small amounts of iron, tin and other alloy metals. In particular, Zircaloy-2 contains about 1.5% tin, 0.15% iron, 0.1% chromium, 0.05% nickel and 0.1% oxygen, whereas Zircaloy-4 contains substantially no nickel and about 0.2% iron but otherwise is similar to Zircaloy-2.
Within the confines of a sealed fuel rod, however, the hydrogen gas generated by the slow reaction between the cladding and residual water may build up to levels which, under certain conditions, can result in localized hydriding of the alloy with concurrent deterioration in the mechanical properties of the alloy. Zircaloy cladding is also adversely affected by such gases as oxygen, nitrogen, carbon monoxide and carbon dioxide at all reactor operating temperatures. These gases react with Zircaloy and other alloys to produce corrosion which can compromise the integrity of the cladding over long service times.
The Zircaloy cladding of a nuclear fuel element is exposed to one or more of the aforementioned gases during irradiation in a nuclear reactor. Sintered refractory and ceramic compositions, such as uranium dioxide and others used as nuclear fuel, release measurable quantities of the aforementioned gases upon heating, such as during fuel element manufacture or especially during irradiation. The reaction of these gases with Zircaloy can result in embrittlement of the cladding which endangers the integrity of the fuel element. Although water and water vapor may not react directly to produce this result, at high temperatures water vapor does react with zirconium and zirconium alloys to produce hydrogen and this gas further reacts locally with the zirconium and zirconium alloys to cause embrittlement.
In light of the foregoing, it is desirable to eliminate, as far as possible, water, water vapor and gases reactive with Zircaloy from the ambient atmosphere during manufacture of cladding. It is also desirable to remove residual contaminants, for example, hydrocarbon-based lubricant, from the surfaces of Zircaloy cladding, which residue can lead to localized corrosion. This is equally true for other Zircaloy components of a nuclear reactor, such as fuel channels, as well as for components made of alloys which are not zirconium-based.